Plasma wall interactions in ITER

Designing the interface between a thermonuclear plasma and the solid material environment is arguably one of the highest technical challenges of ITER and of the successful development of future fusion power reactors. Carbon-based materials are recognized to have superior thermomechanical properties and do not melt, but they retain high levels of tritium by co-deposition with eroded carbon that could severely constrain plasma operations. A carbon-fibre-composite (CFC) divertor target in ITER would survive long enough only with methods of mitigation/suppression of large edge localised modes (ELMs). Metallic materials, such as tungsten, would avoid the tritium retention issue, but melt layer losses due to large ELMs and disruptions may lead to severe damage and unacceptably short lifetimes. Maintaining plasma purity with high-Z materials remains a concern. A mix of several different plasma-facing materials is proposed in ITER to optimize the requirements of areas with different power and particle flux characteristics (i.e., Be for the first wall, CFC for the divertor strike point tiles and W elsewhere in the divertor). However, this will inevitably lead to cross-material contamination and the formation of material mixtures, whose behaviour remains uncertain and requires further investigation. Some of the most outstanding plasma-wall interaction problems that are still at issue in the design of ITER are briefly discussed in this paper, together with prospects for their resolution, by either R&D or design changes and/or dedicated operation provisions. They include: (i) plasma thermal loads and material damage during type I ELMs and disruptions, and prospects for mitigation, (ii) control of co-deposited tritium inventory in the carbon films, (iii) effects of mixed-materials and (iv) plasma operation with tungsten plasma-facing components. Directions and priorities of future research are proposed to narrow the remaining uncertainties and improve confidence in solutions in these areas.